ML20211M265

Amend 113 to DPR-65,changing Tech Specs to Provide for Revised Temp/Pressure Limits in Tech Spec 3/4.4.9, Pressure /Temp Limits & Tech Spec Figure 3.4-2, RCS Pressure Temp Limitations for 12 Full Power Yrs

ML20211M265
Person / Time
Site:	Millstone
Issue date:	12/08/1986
From:	Thadani A Office of Nuclear Reactor Regulation
To:	Connecticut Light & Power Co, Northeast Nuclear Energy Co (NNECO), Western Massachusetts Electric Co
Shared Package
ML20211M272	List: ML20211M265 ML20211M284
References
DPR-65-A-113, TAC 63133, TAC 63198 NUDOCS 8612170091
Download: ML20211M265 (19)

Text

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UNITED STATES

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g NUCLEAR REGULATORY COMMISSION 3

E WASHINGTON, D. C. 20555

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NORTHEAST NUCLEAR ENERGY COMPANY THE CONNECTICUT LIGHT AND POWER COMPANY THE WESTERN MASSACHUSETTS ELECTRIC COMPANY DOCKET NO. 50-336 MILLSTONE NUCLEAR POWER STATION, UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.113-License No. DPR-65

.1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The applications for amendments by Northeast Nuclear Energy Company, et al. (the licensee), dated October 20, October 24, and October 27, 1986 comply with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; i

B.

The facility will opar~.te in conformity with the applications, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

h ADO P

, 2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-65 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No.113, are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

/

S Asho C. Thadani, Director PWR Project Directorate #8 Di ision of PWR Licensing-B

Attachment:

Changes to the Technical Specifications Date of Issuance: December 8, 1986 S

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ATTACHMENT TO LICENSE AMENDMENT NO.113 FACILITY CPERATING LICENSE NO. DPR-65 DOCKET NO. 50-336 Replace the folidwing pages of the Appendix A Technical Specifications with the enclosed pages. The revised pages are identified by amendment number and contain vertical lines indicating the areas of change. The corresponding overleaf. pages are provided to mLintain document completeness.

Remove Page Insert Page 2-2 2-2 2-4 2-4 3/4 2-8(a) 3/4 2-8(a) 3/4 2-9 3/4 2-9 3/4 2-13 3/4 2-13 3/4 2-14 3/4 2-14 3/4 4-17 3/4 4-17 3/4 4-19 3/4 4-19 B3/4 4-6 B3/4 4-6 1

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2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS

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REACTOR CORE 2.1.1 The combination of THERMAL POWER, pressurizer pressure, and maxi-mum cold leg coolant temperature shall not exceed the limits shown on Figure 2.1-1.

APPLICABILITY: MODES 1 and 2.

ACTION:

Whenever the point defined by the combination of maximum cold leg temper-ature and THERMAL POWER has exceeded the appropriate pressurizer pressure line, be in HOT STANDBY within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

REACTOR COOLANT SYSTEM PRESSURE 2.1.2 The Reactor Coolant System pressure shall not exceed 2750 psia.

APPLICABILITY: MODES 1, 2, 3, 4 and 5.

ACTION:

l MODES 1 and 2 Whenever the Reactor Coolant System pressure has exceeded 2750 psia, be in HOT STANDBY with the Reactor Coolant System pressure within its limit within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

MODES 3,;4and5 Whenever the Reactor Coolant System pressure has exceeded 2750 psia, reduce the Reactor Coolant System pressure to within its limit within 5 minutes.

MILLSTONE - UNIT 2 2-1

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SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.2 LIMITING SAFETY SYSTEM SETTINGS REACTOR TRIP SETPOINTS 2.2.1 The reactor protective instrumentation setpoints shall be set consistent with the Trip Setpoint values shown in Table 2.2-1.

APPLICABILITY: AS SHOWN FOR EACH CHANNEL IN TABLE 3.3-1.

ACTION:

With a reactor protective instrumentation setpoint less conservative than the value shown in the Allowable Values column of Table 2.2-1, declare the channel inoperable and apply the applicable ACTION statement requirement of Specification 3.3.1.1 until the channel is restored to OPERABLE status with its trip setpoint adjusted consistent with the Trip Setpoint value.

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TABLE 2.2-1

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REACTOR PROTECTIVE INSTRUMENTATION TRIP SE1 MINT LIMITS FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES-

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Manual Reactor Trip Not Applicable Not Applicable z

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Power Level-High m

Four Reactor Coolant Pumps

< 9.6% above THERMAL POWER,

< 9.7% above THERMAL POWER, with Operating

~ith a minimum setpoint of a minimum of 1 14.7% of RATED

< 14.6% of RATED THERMAL THERMAL POWER, and a maximum of F0WER, and a maximum of

< 106.7% of RATED THERMAL POWER.

< 106.6% of RATED THERMAL POWER.

3.

Reactor Coolant Flow - Low (1)

Four Reactor Coolant Pumps

> 91.7% of reactor coolant

> 90.1% of reactor coolant flow m

1.

Operating flow with 4 pumps operating *.

with 4 pumps operating *.

4.

Reactor Coolant Pump

> 830 rpm

> 823 rpm Speed - Low 5.

Pressurizer Pressure - High 1 2400 psia 1 2408 psia 6.

Containment Pressure - High 1 4.75 psig 1 5.23 psig i

5 7.

Steam Generator Pressure -

> 500 psia

> 492 psia g

Low (2) (5)

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8.

Steam Generator Water

> 36.0% Water Level - each

> 35.2% Water Level - each steam level - Low (5) steam generator generator b

9.

Local Power Density - High (3)

Trip setpoint adjusted to not Trip setpoint adjusted to not

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exceed the limit lines of exceed the limit lines of y

Figures 2.2-1 and 2.2-2 (4).

Figures 2.2-1 and 2.2-2 (4).

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Design Reactor Coolant flow with 4 pumps operating is 340,000 gom.

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FIGURE 3.2-3a Total Radial Peaking Factor vs Allowable Fraction of Rated Thermal Power l

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MILLSTONE - UNIT 2 3/4 2-8 Amendment No. 99 0

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b FIGURE 3.2-3b Total Radial Peaking Factor vs. Allowable Fraction of RATED THERMAL POWER u

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POWER DISTRIBUTION LIMITS TOTALINTEGRATEDRADIALPEAKINGFACTOR-Ff 4

LIMITING CONDITION FOR OPERATION ThecalculatedvalueofFf,definedasFT, p (1+T ) shall be limited 3.2.3 r

r q

l to < l.537.

APPLICABILITY: MODE 1*.

ACTION:

With Ff > 1.537, within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> either:

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-a.

Reduce THERMAL POWER to bring the combination of THERMAL POWER T

and F to within the limits of Figure 3.'2-3b and withdraw the 7

full length CEAs to or beyond the Long Term Steady State Insertion Limits of Specification 3.1.3.6; or b.

Be in at least HOT STANDBY.

SURVEILLANCE REQUIREMENTS 4.2.3.1 The provisions of Specification 4.0.4 are not applicable.

Ff shall be calculated by the expression Ff = F (1+T ) and F T

4.2.3.2 shall r

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be determined to be within its limit at the following intervals:

l a.

Prior to operation above 70 percent of RATED THERMAL POWER after each fuel loading, b.

At least once per 31 days of accumulated operation in MODE 1, and c.,

Within four hours if the AZIMUTHAL POWER TILT (T ) is > 0.020.

q f

4.2.3.3 F

shall be determined each time a calculation of Ff is required by r

using the incore detectors to obtain a power distribution map with all full length CEAs at or above the Long Term Steady State Insertion Limit for the existing Reactor Coolant Pump Combination.

T 4.2.3.4 T shall be determined each time a calculation of F is required q

7 T

and the value of T used to determine F shall be the measured value of T.

q 7

q

• See Special Text Exception 3.10.2.

MILLSTONE - UNIT 2 3/4 2-9 Amendment No. 38,52,79,90,99,113 L

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POWER DISTRIBUTION LIMITS AZIMUTHAL POWER TILT - T q

LIMITING CONDITION FOR OPERATION 3.2.4 The AZIMUTHAL POWER TILT (T ) shall not exceed 0.02, q

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' APPLICABILITY: MODE 1 above 50% of RATED THERMAL POWER

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ACTION:

a.

With the indicated AZIMUTHAL POWER TILT determined to be > 0.02 but < 0.10, either correct the power tilt within two hours or determine within the next 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and at least once per sub-sequent 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, that the TOTAL PLANAR RADIAL PEAKING FACTOR.(F and the TOTAL INTEGRATED RADIAL PEAKING FACTOR (F ) are within r

the limits of Specifications 3.2.2 and 3.2.3.

b.

With the indicated AZIMUTHAL POWER TILT determined to be > 0.10, operation may proceed for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> provided that the TOTAL T

INTEGRATEDRADIALPEAKINGFACTOR.(F)andTOTALPLANARRADIAL 7

PEAKING FACTOR (Ffy) are within the limits of Specifications 3.2.2 and 3.2.3.

Subsequent operation for the purpose of measurement and to identify the cause of the tilt is allowable provided the THERMAL POWER level is restricted to 4 20% of the i

maximum allowable THERMAL POWER level for the exisElng Reactor Coolant Pump combination.

SURVEILLANCE REQUIREMENT i

4.2.4.1 The provisions of Specification 4.0.4 are not applicable.

4.2.4.2 The AZIMUTHAL POWER TILT shall be determined to be within the limit by:

a.

Calculating the tilt at least once per 7 days when the Channel l

High Deviation Alarm is OPERABLE,

• See Special Test Exception 3.10.2.

MILLSTONE - UNIT 2 3/4 2-10 Amendment No. #, f/, 90

POWER DISTRIBUTION LIMITS DNB MARGIN LIMITING CONDITION FOR OPERATION 3.2.6 The DNB margin shall be preserved by maintaining the cold leg temperature, pressurizer pressure, reactor coolant flow rate, ar.1 AXIAL SHAPE INDEX within the limits specified in Table 3.2-1 and Figure 3.2-4.

APPLICABILITY: MODE 1.

ACTION:

With any of the above parameters exceeding its specified limits, restore the parameter to within its above specified limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERilAL POWER to < 5% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.2.6.1 The cold leg temperature, pressurizer pressure, and AXIAL SHAPE INDEX shall be determined to' be within the limits of Table 3.2-1 and Figure 3.2-4 at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The reactor coolant flow rate shall be determined to be within the limit of Table 3.2-1 at least once per 31 days.

4.2.6.2 The provisions of Specification 4.0.4 are not applicable.

MILLSTONE - UNIT 2 3/4 2-13 Amendment No. 38,90,113

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T TABLE 3.2-1 DNB MARGIN LIMITS Four Reactor Coolant Parameter Pumps Operating Cold Leg Temperature

< 549'F Pressurizer Pressure

> 2225 psia

• Reactor Coolant Flow Rate

> 340,000 gpm AXIAL SHAPE INDEX Figure 3.2-4

• Limit not applicable during either a THERMAL POWER ramp increase in excess of 5% of RATED THERMAL POWER per minute or a THERMAL POWER step increase of greater than 10% of RATED THERMAL POWER.

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MILLSTONE - UNIT 2 3/4 2-14 Amendment No. $$,52.77.99.113 1

REACTOR COOLANT SYSTEM 3/4.4.9 PRESSURE / TEMPERATURE LIMITS REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION 3.4.9.1 The Reactor Coolant System (except the pressurizer) temperature and pressure shall be limited in accordance with the limit-lines shown on Figure 3.4-2 during heatup, cooldown, criticality, and inservice leak and hydrostatic testing with:

A maximum heatup of 20'F in any one hour period with Tavg a.

at or below 110*F, 30*F in any one hour period with Tavg at or below 140'F and above 110'F, and 50*F in any one hour period with Tavg above 140*F.

b.

A maximum cooldown of 80 F in any one hour period with Tavg above 300'F and a maximum cooldown of 30'F in any one hour piriod with Tavg at or below 300*F and above 200 F, and 20*F in any one hour period with Tavg at or below 200*F.

A maximum temperature change of 5'F in any one hour period, c.

during hydrostatic testing operations above system design pressure.

APPLICABILITY: MODES 1, 2*,

3, 4 and 5.

ACTION:

With any of the above limits exceeded, restore the temperature and/or pressure to within the limit within 30 minutes; perform an engineering evaluation to detennine the effects of the out-of-limit condition on the structural integrity of the Reactor Coolant System; determine that l

the Reactor Coolant System remains acceptable for continued operations or be in at least HOT STANDBY witnin the next 6 ~ hours and reduce the RCS Tava and pressure to less than 200'F and 500 psia, respectively, within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

• See Special Test Exception 3.10.3.

MILLSTONE - UNIT 2 3/4 4-17 Amendment No. f5,94,113 i

REACTOR COOLANT SYSTEM SURVEILLANCE' REQUIREMENTS 4.4.9.1 The Reactor Coolant System temperature and pressure shall be a.

determined to be within the limits at least once per hour during system heatup, cooldostn, and inservice leak and hydrostatic testing operations.

b.

The Reactor Coolant System temperature and pressure conditions shall be determined to be to the right of the criticality limit line within 15 minutes prior to making the reactor critical.

The reactor vessel material irradiation surveillance specimens c.

shall be removed and examined, to determine changes in material properties, at the intervals shown in Table 4.4-3.

The results of these examinations shall be used to update Figure 3.4-2.

4 MILLSTONE - UNIT 2 3/4 418 m

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3, MILLSTONE - UNIT 2 3/4 4-19 Amendment No. 29.94.113

TABLE 4.4-3 REACTOR VESSEL MATERIAL IRRADIATION SURVEILLANCE SCHEDULE CAPSULE SCHEDULE (EFPY)

W-97 3.0 W-104 10.0 W-284 17.0 W-263 24.0 W-277 32.0 W-83 Spare W-97 (Flux Monitor) 10.0 i

MILLSTONE - UNIT 2 3/44-20 Amendment No. 94 l

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REACTOR COOLANT SYSTEM BASES Reducing T to < 515*F prevents the release of activity should a steamgeneratorI0berupturesincethesaturationpressureofthe a

primary coolant is below the lift pressure of the atmospheric steam relief valves. The surveillance requirements provide adequate assurance that excessive specific activity levels in the primary coolant will be detected in sufficient time to take corrective action.

Information obtained on iodine spiking will be used to assess the parameters associated with iodine spiking phenomena. A reduction in frequency of isotopic analyses following power changes may be permissible if justified by the data obtained.

I 3/4.4.9 PRESSURE / TEMPERATURE LIMITS All components in the Reactor Coolant System are designed to with-stand the effects of cyclic loads due to system temperature and pressure i

changes. These cyclic loads are introduced by normal load transients, reactor trips, and startup and shutdown operations. The various categories of load cycles used for design purposes are provided in Section 4.0 i

of the FSAR. During startup and shutdown, the rates of temperature and j

pressure changes are limited so that the maximum specified heatup and i

cooldown rates are consistent with the design assumptions and satisfy the stress limits for cyclic operation.

During heatup, the thermal gradients in the reactor vessel wall produce thermal stresses which vary from compressive at the inner wall to tensile at the outer wall. These thermal induced compressive stresses tend to alleviate the tensile stresses induced by the internal pressure.

Therefore, a pressure-temperature curve based on steady state conditions (i.e., no thermal stresses) represents a lower bound of all similar curves for finite heatup rates when the inner wall of the vessel is j

-t treated as the governing location.

t The heatup analysis also covers the determination of pressure-i temperature limitations for the case in which the outer wall of the i

vessel becomes the controlling location. The thermal gradients estab-J 11shed during heatup produce tensile stresses at the outer wall of the vessel. These stresses are additive to the pressure induced tensile stresses which are already present. The thermal induced stresses at the outer wall of the vessel are tensile and are dependent on both the rate of heatup and the time along the heatup ramp; therefore, a lower bound curve similar to that described for the heatup of the inner wall cannot be defined. Subsequently, for the cases in which the outer wall of the i

vessel becomes the stress controlling location, each heatup rate of i

in.terest must be analyzed on an individual basis.

1 MILLSTONE - UNIT 2 8 3/4 4-5 4

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b REACTOR COOLANT SYSTEM BASES The heatup and cooldown limit curves (Figure 3.4-2) are composite F

curves which were prepared by determining the most conservative case, with -

i either the inside or outside wall controlling, for any heatup or cooldown i

rates of up to the maximums described in Section 3.4.9.1.

The heatup and I

cooldown curves were prepared based upon the most limiting value of the

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predicted adjusted reference temperature at the end of the service period indicated on Figure 3.4-2.

The reactor vessel materials have been tested to determine their initial RTNDT; the results of these tests are shown in Table 4.6-1 of the Final Safety Analysis Report. Reactor operation and resultant fast neutron irradiation will cause an increase in the RTND Therefore, an adjusted i

reference temperature, based upon the fluence,T.can be predicted using the methods described in SECY-82-465, "NRC Staff Evaluation of Pressurized Thermal Shock," N)vember 1982. Because it is more conservative, this method was used rather than the proposed Revision 2 to Regulatory Guide 1.99.

The heatup and cooldown limit curves shown on Figure 3.4-2 include predicted adjustments for this shift in RT st the end of the applicable l

service period, as well as adjustments for kssible errors in the pressure H

and temperature sensing instruments.

The actual shift in RT periodicallyduringoperati$kTofthevesselmaterialwillbeestablished by removing and evaluating, in accordance with ASTM E185-73, reactor vessel material irradiation surveillance speci-mens installed near the inside wall of the reactor vessel in the core area.

i Since the neutrun spectra at the irradiation samples and vessel inside l

radius are essentially identical, the measured transition shift for a i

sample can be applied with confidence to the adjacent section of the i

reactor vessel. The heatop and cooldown curves must be recalculated when the ART determined from the surveillance capsule is different from the calcula!SIART for the equivalent capsule radiation exposure.

NDT

'The pressure-temperature limit lines shown on Figure 3.4-2 for reactor i

criticality and for inservice leak and hydrostatic testing have been

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provided to assure compliance with the minimum temperature requirements i

of Appendix G to 10 CFR 50 for reactor criticality and for inservice leak and hydrostatic testing, i

The maximum RT nT for all reactor coolant system pressure-retaining N

materials, with the Exception of the reactor pressure vessel, has been determined to be 50'F. The Lowest Service Temperature limit line t

shown on Figure 3.4-2 is based upon this RT since Article N8-2332 L

(Summer Addenda of 1972) of Section III of VRI ASME Boiler and Pressure j

Vessel Code requires the Lowest Service Temperature to the RTNDT + 100'F L

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j MILLSTONE - UNIT 2 B 3/4 4-6 Amendment No. 94,113

- _ _